Low Q95 Research Articles

Ti/Te effects on transport in EAST low q95 plasmas

Enhanced confinement is observed in EAST low q95 ( q95<3.5 ) plasmas with the temperature ratio ( Ti/Te ) increasing. A statistically significant positive correlation between H98y2 and Ti/Te has been established. Increased Ti triggerd by Neutral Beam Injection in experiments maintaining constant Te leads initially to an improved Ti/Te ratio, which subsequently mitigates heat transport in the ion channel. The Trapped Gyro–Landau Fluid transport model was applied to analyse the dependence of the turbulent transport on Ti/Te . Simulation results demonstrate that the Ion Temperature Gradient mode is stabilized in high Ti/Te plasma, with ion flux results from predictive modelling further validating this effect. These findings corroborate the Weiland model and complement one another. This study contributes to a deeper understanding of how the dimensionless parameter Ti/Te influences plasma transport, and helps demonstrate the feasibility of operations under the low q95 condition required for the ITER-inductive scenario in the new Ti/Te regime.

Experimental and simulation studies on ELM characteristics under a plasma current ramping up discharge on EAST

Previous experimental results show that the poloidal mode spacing of the filamentary structures increases and the dominant toroidal mode number decreases in the edgelocalized mode (ELM) rising phase with increasing plasma current. In addition, the experimental results in this paper show that the energy loss ratio of the pedestal (ΔW/Wped) decreases as the edge safety factor (q95) increases. The BOUT++ three-field two-fluid model can reproduce the experimental results and provide a possible explanation mechanism. The pedestal density plays an important role in the characteristics of filamentary structures as the current ramps up. On the one hand, the resistivity related to the pedestal density drives the instability of the peeling–ballooning mode, and the resistive effect is stronger in the high current case, making the dominant toroidal mode number lower and the corresponding poloidal mode spacing wider in the high current case. A low q95 corresponds to a high pedestal collision rate and a high pedestal energy loss ratio. On the other hand, the ELM crash process is dominated by resistivity, so the ratio of pedestal energy loss caused by ELM is not inversely proportional to the pedestal collision rate.

The onset distribution of rotating m,n=2,1 tearing modes and its consequences on the stability of high-confinement-mode plasmas in DIII-D

Analysis of a multi-scenario database of over 13 000 DIII-D H-mode discharges shows that the m,n=2,1 magnetic islands are dominantly pressure gradient driven, stochastically triggered non-linear instabilities at all edge safety factor ( q95 ) values. The instability onset time closely follows the exponential distribution in intermediate and high q95 scenarios and is characterized by near constant onset rate (λ), in accordance with Poisson-point processes. This implies that the plasmas are operated in marginally stable conditions, characterized by a small threshold for instability growth and variations in the trigger amplitude and/or the stabilizing mechanisms with temporally uniform random distribution in this database. While the majority of the tearing modes occur in the first current-profile relaxation time of the βN flattop, constant λ throughout the βN flattop shows that the tearing onset is insensitive to the evolution of the equilibrium current profile. In low q95 scenarios, where a large fraction of the plasmas are operated at low torque, λ increases over the course of the βN flattop, showing that these plasmas evolve toward more unstable conditions. The onset rate rapidly increases with βN , while it does not show a clear dependence on the current gradient at the mode rational surface. Overall, these observations support that the majority of the analyzed 2,1 tearing modes are non-linear, neoclassically driven instabilities and classical stability does not play a dominant role in their onset.

ECRH system upgrade design using dual frequency gyrotrons for EAST

The ECRH (Electron Cyclotron Resonance Heating) system was used in the EAST (Experimental Advanced Superconducting Tokamak) experiments for several years, and some good experimental results have been obtained. However, the performance of the gyrotrons is degraded and the ability to stabilize output RF (Radio Frequency) power is reduced. Two single-frequency 140 GHz gyrotrons are planned to be repaired in order to improve the performance of the gyrotrons. At the same time, some internal components of the gyrotrons (such as collectors, electron guns, mode conversion mirrors) will be upgraded to make the gyrotrons have the output capability of 0.6 MW/1 MW at 105 GHz/140 GHz. The MOUs (Matching Optical Units), transmission lines, the low voltage power supplies, the measurement and control system, and the water-cooling system also need to be modified for the dual-frequency operation of the new gyrotrons. The retrofitted design of the EAST ECRH system using dual-frequency gyrotrons is discussed in the paper. The upgraded ECRH system will be used for future EAST experiments to explore the range of low q95 and high βN, to study the steady-state operation mode for the future nuclear fusion power plants.

Sustained radiative divertor in high-performance grassy-ELM regime with metal wall towards long-pulse operation in EAST

Abstract Simultaneous control of transient heat load induced by large-amplitude edge-localized modes (ELMs) and steady-state heat load on divertor targets under metal wall environment is crucial for steady-state operation of future tokamak fusion reactors, such as ITER and the China Fusion Engineering Test Reactor (CFETR). In the recent experiments, sustained partial energy detachment without confinement degradation has been achieved in the Experimental Advanced Superconducting Tokamak (EAST) in high-performance grassy-ELM H-mode with q95 ~ 5.9 by a newly developed detachment feedback control scheme, in which we first used electron temperature (Tet) measured by divertor Langmuir probes to identify the onset of energy detachment, and then the system switched to the feedback control of total radiation power measured by absolute extreme ultraviolet (AXUV) system. Tet around the upper outer strike point was successfully maintained less than 8 eV with seeding of 80% Ne and 20% D2 mixture from upper outer divertor, and the total radiation power was maintained ~1.4 MW, around 52% of injected power. There was no significant decrease of the plasma stored energy and H98,y2 factor (~1) over the entire detachment feedback control process. These experiment results demonstrate good compatibility of the high-performance grassy-ELM regime with radiative divertor. In order to confirm the compatibility in a wider range, stable partial energy detachment in grassy-ELM H-mode with relatively lower q95 (~5.4) was also achieved in EAST through the newly developed integrated-feedback-control technique. The new detachment feedback control without confinement degradation in grassy-ELM H-mode provides a candidate mode for EAST long-pulse operation in the future with well control of ELM-induced transient and steady heat fluxes on the divertor target.

Divertor detachment with neon seeding in grassy-ELM H-mode in EAST

Simultaneous control of the transient heat load induced by the large-amplitude edge-localized modes (ELMs) and the time-averaged heat and particle fluxes to the divertor targets is a critical issue for the steady-state operation of a future tokamak fusion reactor. The combination of divertor detachment and grassy-ELM regime provides a candidate solution to this issue. The strong particle exhaust capability of the grassy-ELM regime greatly facilitates the achievement of steady-state operation of a detached plasma with divertor impurity seeding. Stable complete detachment at inner target and partial detachment at outer target in the grassy-ELM regime have been achieved with seeding of 5% neon (Ne) and 95% D2 mixture since 2018 in the EAST superconducting tokamak with ITER-like tungsten monoblock upper divertor. The peak ion fluxes (Γion) on the upper outer and inner divertor targets were reduced by 55% and 92%, respectively. However, it was accompanied by confinement degradation with 18% reduction in the plasma stored energy WMHD. In contrast, partial detachment on the upper outer divertor target with confinement improvement has been achieved with the power across the separatrix (Psep) around the H-mode threshold power. In addition, at relatively lower q95 (∼5.7) with unfavorable Bt direction, Ne seeding leads to a transition from mixed (large/small) ELM regime into grassy-ELM regime with significantly increased ELM frequency, which extends the accessible parameter range of the grassy-ELM regime towards lower q95.

The dominant micro-turbulence instabilities in the lower q95 high βp plasmas on DIII-D and predict-first extrapolation

Large-radius internal transport barriers (ITB) are the signature of high scenarios on DIII-D. Previous studies show that a large Shafranov shift, rather the shear, suppresses the turbulence and helps in the formation of the large-radius ITB. New gyrokinetic simulations suggest that the remaining micro-instabilities in lower q95 (<7.0), high ITB plasmas are drift wave instabilities, including the collisionless trapped electron mode in the core and ITB peak gradient region, the electron temperature gradient mode in the ITB peak gradient region and at the ITB foot and the ion temperature gradient mode at the ITB foot. Gyrokinetic simulation results qualitatively agree with the density fluctuation analysis from beam emission spectroscopy, which suggests the existence of a low-k ion mode at the ITB foot. The gyrokinetic simulations also predict that a larger Shafranov shift can overwhelm the driving sources for turbulence from the profile gradients at higher , leading to stronger turbulence suppression and stronger ITBs in lower q95, high plasmas.

Test of the ITER-like resonant magnetic perturbation configurations for edge-localized mode crash suppression on KSTAR

KSTAR has demonstrated divertor heat flux broadening during edge-localized mode (ELM) crash suppression using ITER-like three-row resonant magnetic perturbation (RMP) for the first time. To address a couple of critical issues in ITER RMP, robust ELM-crash-suppression methodology has been explored at low q95 and established in KSTAR using low-n RMPs. Taking full advantage of the ITER-like three-row in-vessel control coils (IVCC) in KSTAR, a set of intentionally misaligned RMP configurations (IMC) was tested to investigate whether or not IMC could be compatible with ELM crash suppression, while minimizing electromagnetic loads on RMP coils. As a result, the ITER-like three-row IMCs were found not only to have been compatible with the ELM crash suppression, but also to have broadened the divertor heat fluxes in the vicinity of the outer strike point. In comparison, the two-row RMPs have rarely affected the near scrape-off layer (SOL) heat flux despite slightly broadened profile change in the far-SOL. Since the divertor heat flux broadening reflects the dispersal of the peaked near-SOL heat flux, the experimental outcome is quite favorable to the ITER choice of three rows, instead of two rows. Nonetheless, since the IMC-driven broadening observed in the attached plasmas in KSTAR might appear substantially different in the partially detached plasmas in ITER, additional investigation has been conducted to see if RMP-driven, ELM crash suppression could be compatible with detached plasmas. Although no detached plasmas have been identified with ELM crash suppression yet, significantly reduced divertor heat flux was confirmed in high density, ELM-crash-suppressed plasmas at q95 = 3.4 using n = 2 RMPs. These new findings elevate the confidence level about the ITER RMP system, while the remaining uncertainties need to be further clarified using the three-row IVCCs in KSTAR.

Forecasting electron cyclotron current drive stabilization of neoclassical tearing modes in ITER

The ITER baseline design relies on ECCD to stabilize confinementdegrading disruptive NTMs [1]. However, the EC power required will take a toll on the fusion gain Q. The MDC-8 group (in existence since 2005) has the goal to provide a range of data to benchmark the Rutherford tearing stability equation for NTM evolution, allowing predictions for ITER ECCD requirements to be validated. Experimental contributors have included ASDEX UPGRADE, DIII-D, EAST, FTU, HL-2A, JT-60U, KSTAR, TEXTOR and TCV. While any m/n tearing mode island can reduce confinement, the m=2, n=1 mode at q=2 is particularly damaging. This mode is at a relatively large minor radius in the low q95∼3 safety factor of ITER and thus close to the resistive wall; with the relatively low rotation in ITER (large inertia, small torque), an uncontrolled mode will rapidly lock at low tearing mode amplitude with subsequent disruption [2]. While progress is being made in modeling of the stability space and control [3] and experiments are promising, implementation still needs to be successfully demonstrated experimentally.The ITPA consensus is that ITER's 24 1-MW gyrotrons will provide more than sufficient EC power from the upper launch mirrors to drive narrow (but not too narrow) ECCD at q=2 for stabilization, with good alignment. Broadening of the ECCD, by edge turbulence for example, is a concern that would demand more EC power but also make alignment easier. Pre-emption at lower CW power or active stabilization by early mode onset detection and higher peak (possibly lower average) pulsed power are issues still under continuing investigation.Most EC-NTM experimental studies so far are at relatively high q95 with smaller radius at q=2 and thus higher Te for better current drive efficiency, higher rotation and weaker wall coupling. DIII-D, for example, is now well poised to pursue ECCD NTM stabilization at both low q95 and at low rotation in the 2017 campaign. The MDC-8 as a whole is proceeding to narrow the experimental focus for a comparison of observations from different devices. This will establish the physics basis for successful stabilization in ITER.

Interaction of external n = 1 magnetic fields with the sawtooth instability in low-q RFX-mod and DIII-D tokamaks

External n = 1 magnetic fields are applied in RFX-mod and DIII-D low safety factor Tokamak plasmas to investigate their interaction with the internal MHD dynamics and in particular with the sawtooth instability. In these experiments the applied magnetic fields cause a reduction of both the sawtooth amplitude and period, leading to an overall stabilizing effect on the oscillations. In RFX-mod sawteeth eventually disappear and are replaced by a stationary m = 1, n = 1 helical equilibrium without an increase in disruptivity. However toroidal rotation is significantly reduced in these plasmas, thus it is likely that the sawtooth mitigation in these experiments is due to the combination of the helically deformed core and the reduced rotation. The former effect is qualitatively well reproduced by nonlinear MHD simulations performed with the PIXIE3D code. The results obtained in these RFX-mod experiments motivated similar ones in DIII-D L-mode diverted Tokamak plasmas at low q95. These experiments succeeded in reproducing the sawtooth mitigation with the approach developed in RFX-mod. In DIII-D this effect is correlated with a clear increase of the n = 1 plasma response, that indicates an enhancement of the coupling to the marginally stable n = 1 external kink, as simulations with the linear MHD code IPEC suggest. A significant rotation braking in the plasma core is also observed in DIII-D. Numerical calculations of the neoclassical toroidal viscosity (NTV) carried out with PENT identify this torque as a possible contributor for this effect.

Recent ASDEX Upgrade research in support of ITER and DEMO

Recent experiments on the ASDEX Upgrade tokamak aim at improving the physics base for ITER and DEMO to aid the machine design and prepare efficient operation. Type I edge localized mode (ELM) mitigation using resonant magnetic perturbations (RMPs) has been shown at low pedestal collisionality . In contrast to the previous high ν* regime, suppression only occurs in a narrow RMP spectral window, indicating a resonant process, and a concomitant confinement drop is observed due to a reduction of pedestal top density and electron temperature. Strong evidence is found for the ion heat flux to be the decisive element for the L–H power threshold. A physics based scaling of the density at which the minimum PLH occurs indicates that ITER could take advantage of it to initiate H-mode at lower density than that of the final Q = 10 operational point. Core density fluctuation measurements resolved in radius and wave number show that an increase of R/LTe introduced by off-axis electron cyclotron resonance heating (ECRH) mainly increases the large scale fluctuations. The radial variation of the fluctuation level is in agreement with simulations using the GENE code. Fast particles are shown to undergo classical slowing down in the absence of large scale magnetohydrodynamic (MHD) events and for low heating power, but show signs of anomalous radial redistribution at large heating power, consistent with a broadened off-axis neutral beam current drive current profile under these conditions. Neoclassical tearing mode (NTM) suppression experiments using electron cyclotron current drive (ECCD) with feedback controlled deposition have allowed to test several control strategies for ITER, including automated control of (3,2) and (2,1) NTMs during a single discharge. Disruption mitigation studies using massive gas injection (MGI) can show an increased fuelling efficiency with high field side injection, but a saturation of the fuelling efficiency is observed at high injected mass as needed for runaway electron suppression. Large locked modes can significantly decrease the fuelling efficiency and increase the asymmetry of radiated power during MGI mitigation. Concerning power exhaust, the partially detached ITER divertor scenario has been demonstrated at Psep/R = 10 MW m−1 in ASDEX Upgrade, with a peak time averaged target load around 5 MW m−2, well consistent with the component limits for ITER. Developing this towards DEMO, full detachment was achieved at Psep/R = 7 MW m−1 and stationary discharges with core radiation fraction of the order of DEMO requirements (70% instead of the 30% needed for ITER) were demonstrated. Finally, it remains difficult to establish the standard ITER Q = 10 scenario at low q95 = 3 in the all-tungsten (all-W) ASDEX Upgrade due to the observed poor confinement at low βN. This is mainly due to a degraded pedestal performance and hence investigations at shifting the operational point to higher βN by lowering the current have been started. At higher q95, pedestal performance can be recovered by seeding N2 as well as CD4, which is interpreted as improved pedestal stability due to the decrease of bootstrap current with increasing Zeff. Concerning advanced scenarios, the upgrade of ECRH power has allowed experiments with central ctr-ECCD to modify the q-profile in improved H-mode scenarios, showing an increase in confinement at still good MHD stability with flat elevated q-profiles at values between 1.5 and 2.

Fast-ion redistribution and loss due to edge perturbations in the ASDEX Upgrade, DIII-D and KSTAR tokamaks

The impact of edge localized modes (ELMs) and externally applied resonant and non-resonant magnetic perturbations (MPs) on fast-ion confinement/transport have been investigated in the ASDEX Upgrade (AUG), DIII-D and KSTAR tokamaks. Two phases with respect to the ELM cycle can be clearly distinguished in ELM-induced fast-ion losses. Inter-ELM losses are characterized by a coherent modulation of the plasma density around the separatrix while intra-ELM losses appear as well-defined bursts. In high collisionality plasmas with mitigated ELMs, externally applied MPs have little effect on kinetic profiles, including fast-ions, while a strong impact on kinetic profiles is observed in low-collisionality, low q95 plasmas with resonant and non-resonant MPs. In low-collisionality H-mode plasmas, the large fast-ion filaments observed during ELMs are replaced by a loss of fast-ions with a broad-band frequency and an amplitude of up to an order of magnitude higher than the neutral beam injection prompt loss signal without MPs. A clear synergy in the overall fast-ion transport is observed between MPs and neoclassical tearing modes. Measured fast-ion losses are typically on banana orbits that explore the entire pedestal/scrape-off layer. The fast-ion response to externally applied MPs presented here may be of general interest for the community to better understand the MP field penetration and overall plasma response.

Quiescent H-mode operation using torque from non-axisymmetric, non-resonant magnetic fields

Quiescent H-mode (QH-mode) sustained by magnetic torque from non-axisymmetric magnetic fields is a promising operating mode for future burning plasmas including ITER. Using magnetic torque from n = 3 fields to replace counter-Ip torque from neutral beam injection, we have achieved long duration, counter-rotating QH-mode operation with neutral beam injection (NBI) torque ranging continuously from counter-Ip up to co-Ip values of about 1 N m. This co-Ip torque is about 3 times the scaled torque that ITER will have. This range also includes operation at zero net NBI torque, applicable to rf wave heated plasmas. These n = 3 fields have been created using coils either inside or, most recently, outside the toroidal coils. Experiments utilized an ITER-relevant lower single-null plasma shape and were done with ITER-relevant values , and βN = 2. Discharges have confinement quality H98y2 = 1.3, exceeding the value required for ITER. Initial work with low q95 = 3.4 QH-mode plasmas transiently reached fusion gain values of G = βN , which is the desired value for ITER; the limits on G have not yet been established. This paper also includes the most recent results on QH-mode plasmas run without n = 3 fields and with co-Ip NBI; these shots exhibit co-Ip plasma rotation and require NBI torque ⩾2 N m. The QH-mode work to date has made significant contact with theory. The importance of edge rotational shear is consistent with peeling–ballooning mode theory. We have seen qualitative and quantitative agreement with the predicted torque from neoclassical toroidal viscosity.

Confinement and edge studies towards low ρ* and ν* at JET

The size and capability of JET to reach high plasma current and field enables a study of the plasma behaviour at ion Larmor radius and collisionality values approaching those of ITER. In this paper such study is presented. The achievement of stationary type I ELMy H-modes at high current proved to be quite challenging. As the plasma current was increased, it became more difficult to achieve stationary conditions. Nevertheless, it was possible to achieve stable operation at high plasma current (up to 4.5 MA) and low q95 (2.65–3) at JET. One of the main reasons to revisit the high plasma current experiments done in 1997 is the higher power available and the improvement of the pedestal diagnostics. Indeed, compared with previous results, higher stored energy was achieved but confinement was still degraded. The causes of this confinement degradation are discussed in the paper.

High current and low q95 scenario studies for FAST in the view of ITER and DEMO

The Fusion Advanced Study Torus (FAST) has been proposed as a possible European satellite, in view of ITER and DEMO, in order to: (a) explore plasma wall interaction in reactor relevant conditions, (b) test tools and scenarios for safe and reliable tokamak operation up to the border of stability, and (c) address fusion plasmas with a significant population of fast particles. A new FAST scenario has been designed focusing on low-q operation, at plasma current IP=10MA, toroidal field BT=8.5T, with a q95≈2.3 that would correspond to IP≈20MA in ITER. The flat-top of the discharge can last a couple of seconds (i.e. half the diffusive resistive time and twice the energy confinement time), and is limited by the heating of the toroidal field coils. A preliminary evaluation of the end-of-pulse temperatures and of the electromagnetic forces acting on the central solenoid pack and poloidal field coils has been performed. Moreover, a VDE plasma disruption has been simulated and the maximum total vertical force applied on the vacuum vessel has been estimated.

Reactor-relevant quiescent H-mode operation using torque from non-axisymmetric, non-resonant magnetic fields

Results from recent experiments demonstrate that quiescent H-mode (QH-mode) sustained by magnetic torque from non-axisymmetric magnetic fields is a promising operating mode for future burning plasmas. Using magnetic torque from n=3 fields to replace counter-Ip torque from neutral beam injection (NBI), we have achieved long duration, counter-rotating QH-mode operation with NBI torque ranging from counter-Ip to up to co-Ip values of 1-1.3 Nm. This co-Ip torque is 3 to 4 times the scaled torque that ITER will have. These experiments utilized an ITER-relevant lower single-null plasma shape and were done with ITER-relevant values of νped* and βNped. These discharges exhibited confinement quality H98y2=1.3, in the range required for ITER. In preliminary experiments using n=3 fields only from a coil outside the toroidal coil, QH-mode plasmas with low q95=3.4 have reached fusion gain values of G=βNH89/q952=0.4, which is the desired value for ITER. Shots with the same coil configuration also operated with net zero NBI torque. The limits on G and co-Ip torque have not yet been established for this coil configuration. QH-mode work to has made significant contact with theory. The importance of edge rotational shear is consistent with peeling-ballooning mode theory. Qualitative and quantitative agreements with the predicted neoclassical toroidal viscosity torque is seen.

Hybrid-like 2/1 flux-pumping and magnetic island evolution due to edge localized mode-neoclassical tearing mode coupling in DIII-D

Direct analysis of internal magnetic field pitch angles measured using the motional Stark effect diagnostic shows m/n=2/1 neoclassical tearing modes exhibit stronger poloidal magnetic flux-pumping than typical hybrids containing m/n=3/2 modes. This flux-pumping causes the avoidance of sawteeth, and is present during partial electron cyclotron current drive suppression of the tearing mode. This finding could lead to hybrid discharges with higher normalized fusion performance at lower q95. The degree of edge localized mode-neoclassical tearing mode (ELM-NTM) coupling and the strength of flux-pumping increase with beta and the proximity of the modes to the ELMing pedestal. Flux-pumping appears independent of magnetic island width. Individual ELM-NTM coupling events show a rapid timescale drop in the island width followed by a resistive recovery that is successfully modeled using the modified Rutherford equation. The fast transient drop in island width increases with ELM size.

Overview of ASDEX Upgrade results

The ASDEX Upgrade programme is directed towards physics input to critical elements of the ITER design and the preparation of ITER operation, as well as addressing physics issues for a future DEMO design. After the finalization of the tungsten coating of the plasma facing components, the re-availability of all flywheel-generators allowed high-power operation with up to 20 MW heating power at Ip up to 1.2 MA. Implementation of alternative ECRH schemes (140 GHz O2- and X3-mode) facilitated central heating above ne = 1.2 × 1020 m−3 and low q95 operation at Bt = 1.8 T. Central O2-mode heating was successfully used in high P/R discharges with 20 MW total heating power and divertor load control with nitrogen seeding. Improved energy confinement is obtained with nitrogen seeding both for type-I and type-III ELMy conditions. The main contributor is increased plasma temperature, no significant changes in the density profile have been observed. This behaviour may be explained by higher pedestal temperatures caused by ion dilution in combination with a pressure limited pedestal and hollow nitrogen profiles. Core particle transport simulations with gyrokinetic calculations have been benchmarked by dedicated discharges using variations of the ECRH deposition location. The reaction of normalized electron density gradients to variations of temperature gradients and the Te/Ti ratio could be well reproduced. Doppler reflectometry studies at the L–H transition allowed the disentanglement of the interplay between the oscillatory geodesic acoustic modes, turbulent fluctuations and the mean equilibrium E × B flow in the edge negative Er well region just inside the separatrix. Improved pedestal diagnostics revealed also a refined picture of the pedestal transport in the fully developed H-mode type-I ELM cycle. Impurity ion transport turned out to be neoclassical in between ELMs. Electron and energy transport remain anomalous, but exhibit different recovery time scales after an ELM. After recovery of the pre-ELM profiles, strong fluctuations develop in the gradients of ne and Te. The occurrence of the next ELM cannot be explained by the local current diffusion time scale, since this turns out to be too short. Fast ion losses induced by shear Alfvén eigenmodes have been investigated by time-resolved energy and pitch angle measurements. This allowed the separation of the convective and diffusive loss mechanisms.

Avoidance of disruptions at high βN in ASDEX Upgrade with off-axis ECRH

Experiments on disruption avoidance have been carried out in H-mode ASDEX Upgrade plasmas: the localized perpendicular injection of ECRH (1.5 MW ∼ 0.2Ptot) onto the q = 2 resonant surface has led to the delay and/or complete avoidance of disruptions in a high βN scenario (Ip = 1 MA, Bt = 2.1 T, q95 ∼ 3.6, with NBI ∼7.5 MW). In these discharges (at low q95 and low density) neoclassical tearing modes (NTMs) are excited: the growth and locking of the m/n = 2/1 mode leads to the disruption. The scheme of the experiment is successfully applied in the same way as in previous disruption avoidance experiments in FTU and ASDEX Upgrade. As soon as the disruption precursor signal (the locked mode detector and/or the loop voltage) reaches the preset threshold, the ECRH power is triggered by real-time control. A poloidal scan in deposition location (ρdep) has been carried out by setting the poloidal launching mirrors at different angles in each discharge. The results depend on ρdep: complete disruption avoidance can be achieved when the power is injected close to or onto the 2/1 island. When ECRH is injected outside the island (either at radii inside or outside the q = 2 surface), the discharge is disrupted as in the reference case.

Edge energy transport barrier and turbulence in the I-mode regime on Alcator C-Mod

We report extended studies of the I-mode regime [Whyte et al., Nucl. Fusion 50, 105005 (2010)] obtained in the Alcator C-Mod tokamak [Marmar et al., Fusion Sci. Technol. 51(3), 3261 (2007)]. This regime, usually accessed with unfavorable ion B × ∇B drift, features an edge thermal transport barrier without a strong particle transport barrier. Steady I-modes have now been obtained with favorable B × ∇B drift, by using specific plasma shapes, as well as with unfavorable drift over a wider range of shapes and plasma parameters. With favorable drift, power thresholds are close to the standard scaling for L–H transitions, while with unfavorable drift they are ∼ 1.5–3 times higher, increasing with Ip. Global energy confinement in both drift configurations is comparable to H-mode scalings, while density profiles and impurity confinement are close to those in L-mode. Transport analysis of the edge region shows a decrease in edge χeff, by typically a factor of 3, between L- and I-mode. The decrease correlates with a drop in mid-frequency fluctuations (f ∼ 50–150 kHz) observed on both density and magnetics diagnostics. Edge fluctuations at higher frequencies often increase above L-mode levels, peaking at f ∼ 250 kHz. This weakly coherent mode is clearest and has narrowest width (Δf/f ∼ 0.45) at low q95 and high Tped, up to 1 keV. The Er well in I-mode is intermediate between L- and H-mode and is dominated by the diamagnetic contribution in the impurity radial force balance, without the Vpol shear typical of H-modes.